Ageing of Steam Generators (IAEA TECDOC-981)

181 Pages·1997·13.605 MB·English

IAEA-TECDOC-981
dnAsasessm ent
management of ageing of major
nuclear power plant components
important to safety:
Steam generators
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ASSESSMENT AND MANAGEMENT OF AGEING OF MAJOR
NUCLEAR POWER PLANT COMPONENTS IMPORTANT TO SAFETY:
STEAM GENERATORS
IAEA, VIENNA, 1997
IAEA-TECDOC-981
ISSN 1011-4289
©IAEA, 1997
Printed by the IAEA in Austria
November 1997
FOREWORD
At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA
Member States. Operating experiencs aeh shown that ineffective conte rhot afolg eing
degrade amhttiP oafcPojnooN mr ponents (e.g. causey ubdn anticipated phenoymb ednnaa
operating, maintenance, desr imgona nufacturing ern jreoaorcsp) ardize plant s daanflesatoy
plant life. Ageing in NPPs must be therefore effectively managed to ensure the availability
f doesign functions througe phhloautn tt service life. Fe srhaofmte ty perspective, this means
controlling within acceptable limits the ageing degradation and wearout of plant components
important to safety so that adequate safety margins remain, i.e. integrity and functional
capability in excess of normal operating requirements.
e aa sshesT frrenthoei sd empisniss mo n aoeTrnetaEsnn a gCtoeDm f OsaeoginCe ting
e mhP tcaP ofjNoomr ponents importa onstat fe rehetTyp. oe brrtaas s neeodx periednncae
practices of NPP operators, regulators, designers, manufacturers, and technical support
organizations and a widely accepted Methodology for the Management of Ageing of NPP
Components Importano tSt afety whs iaciwshe s hIutA eybEd A ni1 992. They have been
compiled using contributions from technical experts in typically 10 to 12 countries for each
repora tf, eedback froa mS eptember 1994 Technical Committee Meeting at3te5n dyebd
technical experts from 21 Member States (who reviewed first drafts in specialized working
groups)d nra, eview comments from invited specialists.
The current practices for the assessment of safety margins (fitness-for-service) and the
inspection, monitoring and mitigation of ageing degradation of selected components of Canada
deuterium-uranium (CANDU) reactor, boiling water reactor (BWR), pressurized water reactor
(PWR)d n,wa ater moderated, water cooled energy reactor (WWER) plante srad ocumented
in the reports. These practices are intended to help all involved directly and indirectly in
ensuring the safe operation of NPPs; and also to provide a common technical basis for
dialogue between plant oped rranetgoaurs lators when dealing with age-related licensing
issues. Since the reports are written from a safety perspective, they do not address life or life-
cycle mae nphalgtaen mtf oecno tmponents, which inve oihlnvttees graf toiaogne ing
management and economic planning. The target audience of the reports consists of technical
experts from Nd fnrPoaPms regulatory, plant design, manufac dtutenrcianhgn ical support
organizations dealing with specific plant components addressed in the reports.
The component ade dphrrtee ssnseeind t pube lshictteaa tsimoin geehnter aftoor
CANDU, PWR and WWER nuclear power plants. The contributors to the drafting and review
of this TECDOC are identified at the end of this publication. Their work is greatly
appreciai tephda ,rtice ucholantrt ,ributif oPon.Es . MacDon. aMCldad ,Vrun.sNaka . Shah
are acknowledged. The officer who directed the preparation of the report was J. Pachner of
e Dhtivisif oNon uclear Installation Safety.
EDITORIAI. NOTE
n Ipreparing this publicationr of press, ste hatI ffoAf EA have mae dhtp peu ages fromeht
original manuscript(s). The views expressed do not necessarily reflect those of the IAEA, the
governmen etnsh ootm finating Member en hSottm artioensa ting organizations.
Throughoe huttet xt namef oMs ember Stae trerase taines adth ey were whe ehtsntae wxt
compiled.
The soue f particular designationf osc ountrier ost erritories doet oinsm py njlauy dgemenybt
the e plhuetgba loils ethsIh AtesatErf atos ,uAus ,ch countrr itoeesr ritorf ieoths ,eir authorities
d nianstitutionr e osohd tf elimitation of their boundaries.
The mention of names of specific companies or products (whether or not indicated as
registered) does not imply any intention to infringe proprietary rights, nor should it be construed
as an endorsement or recommendation on the part of the IAEA.
CONTENTS
1. INTRODUCTION ..................................................... 1
1.1. Background ...................................................... 1
1.2. Objective ........................................................ 3
1.3. Scope ........................................................... 3
1.4. Structure ......................................................... 3
2. STEAM GENERATOR DESCRIPTIONS ...............................5...
R reWcir2Pc.1u .lating steam generators ..............................5.... .
2.2. CANDU recirculating steam generators ...............................7.
RonWce-Pt2h.r3o .ugh steam generators .............................4.1.... .
2.4. WWER steam generators ............................................ 16
3. STEAM GENERATOR DESIGN BASIS, FABRICATION AND MATERIALS .... 18
3.1. d sCpnoedcaeif sications .....................................8..1.... .
3.2. Fabrication and materials ............................................ 18
3.2.1. Heat exchanger tubes ........................................ 18
3.2.2. Tube installation in the tubesheet ............................... 21
3.2.3. Tube supports .............................................. 22
3.2.4. Feedwater no dsznhzealel l ...............................7..2...
. 4 STEAM GENERATOR DEGRADATION MECHANISMS ...................82.
4.1. PWR and CANDU recirculating steam generator tubes .................... 28
4.1.1. Primary water stress corrosion cracking (PWSCC) ................. 29
4.1.2. Outside diameter stress corrosion cracking (ODSCC) ............... 36
4.1.3. Fretting, wear and thinning .................................... 38
4.1.4. Pitting .................................................... 41
4.1.5. Denting ................................................... 41
4.1.6. High-cycle fatigue .....................................2.4... .
4.1.7. Wastage .................................................. 43
4.2. PWR once-through steam generator tubes ............................... 43
4.2.1. Erosion-corrosion .....................................4..4... .
4.2.2. High-cycle fatigue .......................................... 44
4.2.3. Low-temperature primary-side stress corrosion cracking ............ 45
4.2.4. Outside diameter intergranular stress corrosion cracking (IGSCC)
and intergranular attack (IGA) ................................54.
4.3. WWER steam generator tubes ........................................ 46
4.4. Tube rupture events ............................................6.4...
4.4.1. Tube ruptures ..........................................6.4...
4.4.2. Incipient tube rupture events .................................. 52
4.d 5naC RW.P ANDU steam generator shell, feedwater nozzled na tubesheet . . . ..25
4.5.1. Corrosion-fatigue ...........................................35
4.5.2. Transgranular stress corrosion cracking .......................... 55
4.5.3. High-cycle fatigue .......................................... 55
4.5.4. Erosion-corrosion ........................................... 55
4.6. WWER collector, shell, and feedwater distribution system .................. 57
4.6.1. Stress corrosion cracke hWitn fgoW ER-1000 collectors .........7.5..
4.6.2. Erosion-core rfoheste idfoown ater distribution system ..........9.5...
4.6.3. Failurf eoc ollector cover bolts ...............................06..
4.7. Summf caouryr rent world experience ............................1.6.... .
5. STEAM GENERATOR AGEING MANAGEMENT:
OPERATIONAL GUIDELINES .......................................... 66
5.1. Primary coolant system water chemistry control parameters ..............6.6..
5.2. Secondary coolant system water chemistry control parameters .........9.6.... .
5.3. Measures to control secondary-side chemical impurity incursions ............ 77
5.4. Measures to remove secondary-side impurities ........................... 79
5.5. Measu ocroetsn trol steam generator deposits ......................0..8... .
6. STEAM GENERATOR INSPECTIOD NMNA ONITORING REQUIREMENTS
AND TECHNOLOGIES ...............................................2.8
6.1. Tubing inspection requirements ...................................2.8...
6.1.1. Tubing inspection requirements in the USA ...................... 82
6.1.2. Tubing inspection requiremen Cnitas nada .................7..8... .
6.1.3. Tubing inspection requiree Cmhztee nnctihs Republic .........8..8...
6.1.4. Tubing inspection requirements in France ........................ 88
6.1.5. Tubing inspection requiremenn tGis ermany ...................88...
6.1.6. Tubing inspection requirements in Japan ......................... 88
6.1.7. Tubing inspection requirements in Russia ........................ 89
6.1.8. Tubing inspection requirements in Slovenia ...................... 89
6.1.9. Tubing inspection requirementsn i Spain .......................9.8.
6.1.10. Tubing inspection requiremen nSitws eden ..................9.8... .
6.1.11. Tubing inspection requirements in Switzerland .................... 90
6.1.12. EPRI tubing inspection recommendations ........................ 90
6.2. Tubing inspection techniques .......................................0.9.
6.2.1. Eddy current testing ......................................... 90
6.2.2. Ultrasonic and other inspection methods ......................... 97
6.2.3. Destructive testing .......................................79...
6.3. Monitoring leakage from tubing ....................................8.9.
6.4. Feedwater nozzle inspection ......................................... 98
6.5. Monitoring fatigue damageo t feedwater nozzles .......................2.01.
. 7 STEAM GENERATOR ASSESSMENT METHODS DNFA ITNESSROF
SERVICE GUIDELINES ............... ..........................5.0..1.. .
7.1. Tubing repair criteria .............. ..............................5.01
7.1.1. Generic fitness-for-service criteria ............................. 105
7.1.2. Defect type and location specific repair criteria ................... 106
7.2. Tubing fitness-for-service guidelines in various countries ................. Ill
7.2.1. Regulatory practices and fitness-for-service guidelines in
t AhSUe. ..............................................llI..
7.2.2. Alternative US fitness-for-service guidelines for outside
diameter IGSCC/IGA at tube support plates ...................... 113
7.2.3. Alternative US fitness-for-service guidelines for PWSCC
in the roll transition region proposed by EPRI ..................... 116
7.2.4. Other alternative fitness-for-service guidelines in the USA ........... 118
7.2.5. Regulatory practd ifcnietans ess-for-service guidelin neBis elgium .8.1.1.
7.2.6. Regulatory practices and fitness-for-service guidelines in Canada ..... 120
7.2.7. Regulatory practicesd na fitness-for-service guidelines
ie nhCt zech Republic ......................................3.21.
7.2.8. Regulatory prd afcinttniceaes ss-for-service guideln iFnriea snce3 .2.1... .
7.2.9. Regulatory practices and fitness-for-service guidelines in Germany .... 125
7.2.10. Regulatory practicd feniast ness-for-service guidelinen Jsi apan ....5.21..
7.2.11. Regulatory practice dnfsai tness-for-service guidelinen sRi ussia ....62.1.
7.2.12. Regulatory practices and fitness-for-service guidelines in Slovenia .... 126
7.2.13. Regulatory practices and fitness-for-service guidelines in Spain ....... 127
7.2.14. Regulatory practices and fitness-for-service guidelines in Sweden ..... 128
7.2.15. Fitness-for-service guidelines in Switzerland ...................... 129
7.3. Statistical methodsr odf egradation growth assessment ....................92.1
7.4. Use of probabilistic fracture mechanics analysis to assess
steam generator ageing .........................................2.3.1.. .
8. STEAM GENERATOR MAINTENANCE: MITIGATION, REPAIR
AND REPLACEMENT ................................................. 135
8.1. Primary-side mitigation techniques for tubes ............................ 135
8.1.1. Rotopeening and shot peening ................................. 135
8.1.2. Stress relieving ............................................5.31
8.1.3. Reduce ihnhogt t-leg side temperature .......................7.3.1..
8.2. Secondary-side mitigation techniqur etosuf bes .......................7.31..
8.3. Tube repair ....................................................... 138
8.3.1. Plugging .................................................. 138
8.3.2. Sleeving .................................................. 140
8.3.3. Nickel plating .............................................. 146
8.4. Vibration control .................................................. 146
8.4.1. Preheater repairs .....................................6..4..1.. .
8.4.2. Antivibratir oarbne placement ..............................7.41..
8.5. Mitigatie hotth fnoe rmal fatig ehfute feoe dwater nozzled npsa iping .......74.1.
8.6. Repair of feedwater nozzles and piping ................................. 148
8.7. Steam generator replacement .....................................9.4.1..
RreWp8l.Pa7c.1em. ent steam generator designs ..................0.5..1..
8.7.2. CANDU replacement designs ...............................25.1.
8.7.3. WWER replacement designs .................................. 152
. S9TEAM GENERATOR AGEING MANAGEMENT PROGRAMME .......3..5.1.. .
9.1. Understanding steam generator ageing and feedback of operating experience ... 155
9.2. Definition of steam generator AMP ................................... 155
9.3. Steam generator operation ...........................................651
9.4. Inspection, monitoring and assessment ................................. 157
9.4.1. Inspecd mtniooann itoring ...............................7.5.1.. .
9.4.2. Assessment ..........................................8..5.1.. .
9.5. Maintenance: mitigation, repair and replacement ......................... 158
REFERENCES .......................................................... 161
ABBREVIATIONS ....................................................... 172
CONTRIBUTO ODRTRS AFD RTENINVAIGE W ........................3.7.1.. .
i.
1.1. BACKGROUND
Managin ehtgs afety aspecf tos nuclear power plant (NPP) ageing requires
implementation of effective programmes for the timely detection and mitigation of ageing
degradation of plant systems, structures and components (SSCs) important to safety, so as to
ensure their integd frunitnay ctional capability throughout plant service life. General guidance
on NPP activities relevant to the management of ageing (maintenance, testing, examination
and inspection of SSCs) is given in the IAEA Nuclear Safety Standards (NUSS) Code on the
Safetyf o Nuclear Power Plants: Operation (Safety Series. oN 50-C-O, Rev.d na )1 associated
Safety Guides on in-service inspection (50-SG-O2), maintenance (50-SG-O7, Rev.l) and
surveillance (50-SG-O8, Rev.l).
Trie Operation Code requires that NPP operating organizations prepare and carry out a
programmef o periodic maintenance, testing, examinatiod nnai nspectionf o plant systems,
structures and components important to safety to ensure that their level of reliability and
effectiveness remains in accord with the design assumptions and intent and that the safety
status of the plant has not been adversely affected since the commencement of operation. This
prog ortaatmk esm iine to ace coohpuetnr tationald clionmnadiy tois ttinhoenars ,applicable
regulatory requiremee nb rtdsena,- evaluae thetl indig hf ot operating experienceehT.
associated Safety Guides provide further guidance on NPP programmes and activities that
contri obtuitmte ely ded tmencittaiigoaf n atoigoen ing degradf aStoiSoCn s impoortatn t
safety.
The Safety Guidn Ioen -Service Inspection (ISI) provides recommendationn mos ethods,
freque dnnacadym inistrativeI S pImr oeeghartsa urmorer fmosc frei tical systdenmas
compone ephntr ftioms ary reactor coolant system aimet add etecting possible deterioration
due to the influences of stress, temperature, irradiation, etc. and at determining whether they
are acceptab roclfeo ntinued safe opere aphttli aofr noonw t hether remedial measuerreas
needed. Organizational and procedural aspects of establishing and implementing an NPP
programme of preventive and remedial maintenance to achieve design performance throughout
the operational life of the plant are covered in the Maintenance Safety Guide. Guidance and
recommendations on surveillance activities, for SSCs important to safety, (i.e. monitoring
plant parameters and systems status, checking and calibrating instrumentation, testing and
inspecting d SeSnvCaalsu ,ating resf utohltes se active ie ptSrirheaoustvr) vindeeiildl ance
Safety Guide. The aim of the surveillance activities is to verify that the plant is operated
withine hpt rescribed operational limitd snca onditionso ,t detectn i timey nda eteriorationfo
SSCs as well as any adverse trend that could lead to an unsafe condition, and to supply data
to be used for assessing the residual life of SSCs. The above Safety Guides provide general
programmatic guidant oncg odei tuv,b e detailed technical advicr eopf articular SSCs.
Ageing management specific programmatic guidance is given in Technical Reports Series
No. 338 "Methodology for the Management of Ageing of Nuclear Power Plant Components
Important to Safety" and in a Safety Practice No. 50-P-3 "Data Collection and Record
Keeping for the Management of Nuclear Power Plant Ageing". Guidance provided in these
reports served as a basis for the development of component specific technical documents